Formation and growth of cracking in feed water pipes and RPV nozzles

Kußmaul, K.; Blind, D.; Jansky, J.; Rintamaa, Rauno

doi:10.1016/0029-5493(84)90256-5

Cited by 20 publications

(9 citation statements)

References 1 publication

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“…[59][60][61][62][63] In most cases, these failures have been associated with thermal loading due to thermal stratification and striping, or with mechanical loading due to vibratory loading. The thermal stratification is caused by the injection of low-flow, relatively cold feedwater during plant startup, hot standby, and variations below 20% of full power, whereas thermal striping is caused by rapid, localized fluctuations of the interface between hot and cold feedwater.…”

Section: Operating Experience In the Nuclear Power Industrymentioning

confidence: 99%

Mechanism and estimation of fatigue crack initiation in austenitic stainless steels in LWR environments.

Chopra¹

2002

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The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figures I-9.1 through I-9.6 of Appendix I to Section III of the Code specify fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Existing fatigue strain-vs.-life (e-N) data illustrate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. This report provides an overview of fatigue crack initiation in austenitic stainless steels in LWR coolant environments. The existing fatigue e-N data have been evaluated to establish the effects of key material, loading, and environmental parameters (such as steel type, strain range, strain rate, temperature, dissolved-oxygen level in water, and flow rate) on the fatigue lives of these steels. Statistical models are presented for estimating the fatigue e-N curves for austenitic stainless steels as a function of the material, loading, and environmental parameters. Two methods for incorporating environmental effects into the ASME Code fatigue evaluations are presented. The influence of reactor environments on the mechanism of fatigue crack initiation in these steels is also discussed. Executive SummarySection III, Subsection NB, of the ASME Boiler and Pressure Vessel Code contains rules for the design of Class 1 components of nuclear power plants. Figures I-9.1 through I-9.6 of Appendix I to Section III specify the Code design fatigue curves for applicable structural materials. However, Section III, Subsection NB-3121, of the Code states that effects of the coolant environment on fatigue resistance of a material were not intended to be addressed in these design curves. Therefore, the effects of environment on fatigue resistance of materials used in operating pressurized water reactor (PWR) and boiling water reactor (BWR) plants, whose primary-coolant pressure boundary components were designed in accordance with the Code, are uncertain.The current Section-III design fatigue curves of the ASME Code were based primarily on strain-controlled fatigue tests of small polished specimens at room temperature in air. Best-fit curves to the experimental test data were first adjusted to account for the effects of mean stress and then lowered by a factor of 2 on stress and 20 on cycles (whichever was more conservative) to obtain the design fatigue curves. These factors are not safety margins but rather adjustment factors that must be applied to experimental data to obtain estimates of the lives of components. They were not intended to address the effects of the coolant environment on fatigue life. Recent fatigue-strain-vs.-life (e-N) data obtained in the U.S. and Japan demonstrate that light water reactor (LWR) environments can have potentially significant effects on the fatigue resistance of materials. Specimen lives obtained from tests in simulated LWR environments can be much shorte...

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Section: Operating Experience In the Nuclear Power Industrymentioning

confidence: 99%

Mechanism and estimation of fatigue crack initiation in austenitic stainless steels in LWR environments.

Chopra¹

2002

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“…22 A fatigue life correction factor F en can be obtained from the statistical model (Eqs. [3][4][5][6][7][8][9], where ln(F en ) = ln(N RTair ) -ln(N water ).…”

Section: Fatigue Life Correction Factormentioning

confidence: 99%

“…Experience with operating nuclear power plants worldwide reveals that many failures may be attributed to fatigue; examples include piping components, nozzles, valves, and pumps. [1][2][3][4][5] In most cases, these failures have been associated with thermal loading due to thermal stratification and striping, or with mechanical loading due to vibration. The effect of these loadings may also have been aggravated by the high-temperature aqueous environments of nuclear power plants.…”

Section: Introductionmentioning

confidence: 99%

Development of a Fatigue Design Curve for Austenitic Stainless Steels in LWR Environments: A Review

Chopra

2002

Pressure Vessel and Piping Codes and Standards

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The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components and specifies fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Existing fatigue strain–vs.–life (ε–N) data illustrate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. This paper reviews the existing fatigue ε–N data for austenitic stainless steels in LWR coolant environments. The effects of key material, loading, and environmental parameters, such as steel type, strain amplitude, strain rate, temperature, dissolved oxygen level in water, and flow rate, on the fatigue lives of these steels are summarized. Statistical models are presented for estimating the fatigue ε–N curves for austenitic stainless steels as a function of the material, loading, and environmental parameters. Two methods for incorporating environmental effects into the ASME Code fatigue evaluations are presented. Data available in the literature have been reviewed to evaluate the conservatism in the existing ASME Code fatigue design curves.

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“…Experience with operating nuclear power plants worldwide reveals that many failures can be attributed to fatigue; examples include piping components, nozzles, valves, and pumps. [1][2][3] In most cases, these failures have been associated with thermal loading that is due to thermal stratification or thermal striping, or with mechanical loading that is due to vibratory loading. Significant thermal loadings due to flow stratification were not included in the original design basis analysis.…”

Section: Environmental Effects On Fatigue Strain-versus-life (S-n) Bementioning

confidence: 99%

Environmentally assisted cracking in light water reactors. Semiannual report, July 1998-December 1998.

Chopra¹,

Chung²,

Gruber³

et al. 1999

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This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1996 to December 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiationassisted stress corrosion cracking of Type 304 SS, IC) EAC of Alloy 600, and (d) characterization of residual stresses in welds of boiling water reactor (BWR) core shrouds by numerical models. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated BWR water at 288°C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRS to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from a low-carbon content heat of Alloy 600 in high-purity oxygenated water at 289°C. Residual stresses and stress intensity factors were calculated for BWR core shroud welds.iii 4. 5. 6. 7.8. 18.19. 20.21. 22.23. 24.25. 26. 27.28. 29.30. 31. 32.33. Plain carbon steels (CSs) and low-alloy steels (LASS) are used extensively in steam supply systems of pressurized and boiling water nuclear reactors (PWRS and BWRs) as piping and pressure vessel materials. Fatigue tests are being conducted on C S s and LASs in water and in air to establish the effects of material and loading variables on fatigue life. The results indicate a significant decrease in fatigue life when five conditions are satisfied simultaneously, viz.. when applied strain range, service temperature, dissolved-oxygen (DO) in the water, and s u l h r content of the steel are above a minimum threshold level, and the loading strain rate is below a threshold value. Only a moderate decrease in fatigue life is observed in light water reactor (LWR) environments when any one threshold condition is not satisfied. This report presents (a) results of fatigue tests that examine the influence of the reactor environment on crack initiation and crack growth of C S s and LASs, and (b) data on austenitic stainless steel (SS) that establish the effects of various loading and environmental variables on fatigue lives of these steels.Crack lengths as a function of fatigue cycles were determined for A533-Gr B LAS and A106-Gr B CS in air and water environments. The results indicate that, in air, fatigue cracks that are 10 p m or longer, form quite early during fatigue (

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Formation and growth of cracking in feed water pipes and RPV nozzles

Cited by 20 publications

References 1 publication

Mechanism and estimation of fatigue crack initiation in austenitic stainless steels in LWR environments.

Mechanism and estimation of fatigue crack initiation in austenitic stainless steels in LWR environments.

Development of a Fatigue Design Curve for Austenitic Stainless Steels in LWR Environments: A Review

Environmentally assisted cracking in light water reactors. Semiannual report, July 1998-December 1998.

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